ML20023C794
| ML20023C794 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 05/05/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20023C791 | List: |
| References | |
| NUDOCS 8305180020 | |
| Download: ML20023C794 (10) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDENT NO.16 TO FACILITY OPERATING LICENSE NPF-12
' SOUTH CAROLINA ELECTRIC & GAS C0!9ANY SOUTH CAROLINA'PUBLIC SERVICE AUTHORITY I.
INTRODUCTION By letters dated February 14,' April 13, (two letters), May 3, and May 4,1983,
- South Carolina Electric and Gas Company submitted specific actions and surveillance programs relative to the V. C. Summer Nuclear Station Model D3 steam generator modification. The proposed surveillance and monitoring program is based on the recommendations made by the Design Review Panel (DRP) in their' January 1983 report, "D2/D3 Steam Generator Design Modification." The staff evaluation (NUREG-0966,.
March 1983) of-the DRP report concluded that the modification of the D2/D3 steam generators is acceptable and that the modified steam generators can be operated at 100% of their design capacity.
The DRP identified three specific items to be addressed by each of the utility owners installing the proposed preheater modifications. These items are as fol-lows:
1.
Provisions should be made for initial monitoring of inlet pressure oscillations; 2.
Plant-specific provisions for assuring feedwater flow and/or feed-water temperature restrictions are met should be described, where applicable; l
3.
Inservice inspection, eddy current testing and tube vibration moni-toring programs and schedules should be described, where applicable.
l The means by which each of the above items-will be implemented are described in i-the licensee's submittals.
~II.
DISCUSSION A.
Feedwater System Changes The V. C. Summer steam generators are provided with separate inlet con-nections for main feedwater and auxiliary feedwater piping. The auxiliary t
feedwater system (AFWS) is used to provide makeup to the steam generators during plant startup until steam generator flow requirements approach the 8305180020 830505 PDR ADOCK 05000395 P
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1 AFWS design capability. At this stage of plant wamup, the main feed-water system istactuated. When main feedwater is introduced into the t
lower main feedwater inlat nozzles, the colder water that has stagnated downstream.of the main feedwater isolation valve is injected into the i
steam generator. Westinghouse has calculated that the ensuing themal transient could result in an overstressed condition of the modification inlet distribution manifold bolts. The licensee proposed modifications to the main feedwater system to prevent excessive themal stresses in the distribution manifold bolts.
In order to eliminate the cold feedwater transient on. the manifold bolts, the licensee has provided a reverse flushing scheme from the steam generator back to the feedwater isolation valve and forward flushing up to the feedwater isolation valve. This will assure a minimum feedwater temperature of 200*F as recommended by Westinghouse for no miniense flow rate restriction. Backflow (reverse) flushing and forward flushing of the main feedwater lines will be performed during plant wamup when the AFWS is being used for steam generator makeup prior to opening of the main feedwater isolation valve. Hot t
i effluent from the steam generator will preheat the main feedwater inlet piping up to the main feedwater isolation valve outside con-i
.tainment. ' Backflow will continue until the main' feedwater piping F
is ' adequately preheated as determined by temperature indicators provided inboard,of the containment' isolation valve. Two new sec-tions of feedwater piping will be:added to permit the backflow to e
be directed to the ' steam generator blowdown system and the forward L
flow to be directed to the deaerator during plant warmup.' These.
L new' sections of piping will be interconnected with the existing main 4
feedwater -line for each steam generator. The new piping sections
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_ ill be located in the intermediate building (essential portion) j wadjacent to the reactor building with a section (nonessential por-tion) located in the turbine building.
i The reverse flush line penetrates the containment, thereby necessi-(
tating the installation of an isolation valve with a Phase "A" isola-l-
tion signal to conform to General Design Criterion (GDC) 57. The l
change to the Technical Specifications ' involves adding the reverse flush line isolation valves (1678A-FW,' 16788-FW, and 1678C-FW) to i
Table' 3.6-11(Containnent Isolation Yalves).
The staff has evaluated the proposed change to the Technical-Specifi-cations and finds it acceptable:since the isolation valves satisfy -the l_
requirements of GDC 57.
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.. 1 The piping and valves in the intermediate building are safety-related, i
' seismic Category I, Quality Group B and protected from natural phe-nomena, including tornado missiles. Redundant safety-related (seismic Category I) valves are provided between the essential.and nonessential portions of the forward flow piping section. The safety-related por-tion of the piping is located in seismic Category I, flood, and tornado protected structures. Thus, the requirements of General Design Criter-ion 2, " Design Basis for Protection A guidelines of Regulatory Guide 1.29, gainst Natural Phenomena," and the Seismic Classification," Positions C.1 and C.2 are satisfied.
The safety-related portion is separated from the effects of internally generated missiles. As indicated in the May 4,1983 letter, the licensee has performed an evaluation of the above feedwater piping modification against the high energy pipe break methodology as indicated in the pre-sent FSAR which was previously approved by the staff. This analysis did include any additional pipe break locations and the effects of pipe whip, jet impingements, flooding, environmental effects, and
- reaction forces on safety-related equipment in the area and will 3
ensure protection of required. safety-related equipment. Thus, the requirements of General Design Criterion 4 " Environmental and Missile j.
Design Bases," are satisfied.
The feedwater system is not required to transfer heat under accident conditions and, therefore, the requirements of General Design Criteria 45, " Inspection of Cooling Water Systems," and 46, " Testing of Cooling Water Systems " are not applicable.
Redundant isolation valves receive signals to close from corresponding control trains to isolate the backflow and forward flow feedwater. lines on automatic start of the emergency feedwater system. _ Therefore, adequate emergency feedwater flow is provioed in the event of a single failure for accidents previously analyzed. Thus, the requirements of i
General Design Criterion 44, " Cooling Water," are satisfied with
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respect to this feedwater modification.
Based on the above, the staff concludes that the modifications to the feedwater system meet the requirements of General Design Criteria 2, 4, and 44 'with respect to protection against natural phenomena, missile and environmental effects, and assuring feedwater isolation for proper functioning of the emergency feedwate-during accident conditions, meets the guidelines of Regulatory Guide 1.29, Positions C.1 and C.2 with respect to its seismic classification, and meets the applicable acceptance criteria of SRP Section 10.4.7, and is, therefore, acceptable.
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Monitoring Program In the DRP evaluation report the DRP recommended that _each utility develop inspection, testing,'_ and monitoring programs specific' to their plant (s). These programs are designed.to verify the hydraulic perfor-mance of the modification and give early indication of any structural problems with the manifold. The DRP's recommended surveillance program L
included visual.-inspection of the manifold assembly and baseline ECT of the affected first five rows of tubes in the preheater sections F
after manifold. installation and visual and ECT after a 6 month full power operational period.
- 1) Visual Inspection 4
1 The visual inspections proposed by South Carolina Electric and Gas Company follow the DRP's recommendations. The visual inspec-tions are intended to provide an early indication of any unexpec-ted loss of structural integrity. Therefore, a visual-inspection of the accessible areas' of the modified components will be per-fomed.
Inspection access will be through the radiography port in the feedwater piping upstream of the steam generator nozzle.-
The inspection will be-performed using a fiber optics boroscope and will be recorded by videotape or still _ photographs for future reference.. Specific items 'to be inspected include bolts and wel_ds for erosion,- fretting wear, corrosion and cracking. The results of the subsequent inspection will be compared with the as-built e
condition of the manifold. Any questionable or unusual visual indications will-be evaluated to determine the need for' corrective action.
If corrective action is' required, a report detailing the 4
problem and the corrective' action will be submitted to the NRC staff prior.to subsequent power operation. The visual inspection described above will be performed following reassembly of the feed-water piping after modification installation and whenever a scheduled
. Technical Specification eddy current test inspection is performed.
The staff finds-the visual inspection program to be acceptable.
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- 2) Eddy Current Inspection The primary method for assessing the effectiveness of the steam generator modification in reducing the rate of tube wear will be
'multifrequency eddy current testing (ECT) using a wear scar stand-ard. The same ECT methods will be used for testing after modifi-cation installation that were used in previous ECT examinations to-allow comparison of results.- The first five rows (45 to 49) will be examined.
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The above inspection will be performed after completion of the modification on each steam generator. This inspection will serve 4
as the baseline inspection for the modified steam generator. A
- second ECT examination will be performed after the initial period of operation. This second examination will include the Technical-Specification 3% random sample inspection.and the same tubes examined during the initial examination. Subsequent exaainations will be in accordance with Technical Specifications plus-an addi-tional 240 tubes total in the preheater sections of all steam gen-erators. The staff finds the ECT program to be acceptaSle.
- 3) Tube Vibration Monitoring 1.
Tube vibration monitoring, using presently installed accelerometers
- i (if operable) will be conducted at power levels of approximately 0%,
30%, 50%, 65%, 75%, 90% and 100% during the initial power ascension to _100% power and at approximately 100%. power after approximately three effective full power months of operation following modifica-tion completion. The staff finds the tube vibration monitoring pro-gram to be acceptable.
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- 4) Loose' Parts Monitoring l
The DRP's recommended surveillance program did not include the use of loose parts monitoring as one means of assuring the continued structural integrity of the installed manifold.
V. C. Stanmer has an installed louse parts detection system (LPDS). This system includes a sensor on the lower head of each. steam generator, which
- would be used to' detect loose parts from the manifold.
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Technical Specifications require that _ daily channel checks, monthly
~ operational tests and 18 month calibrations be performed on the LPDS. Further, the Technical Specifications require that the LPDS be operable. A report must be submitted to NRC if any channel is inoperable for more than thirty days.
With the above Technical Specification requiremeniis, there is sufficient assurance that the LPDS will be operable and a loose manifold could be detected.
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- 5) Inlet Pressure Monitoring In Section 5.2.13 of its report, the DRP recommended that the pressure oscillations in the feedline be initially monitored throughout the design operating flow range. To accomplish this the licensee has proposed the following pressure monitoring pro-gram.
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Pressure transducers for detection of feedwater pressure pulsations will be installed in each feedwater line adjacent to the steam
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. generator feedwater nozzle. To verify that the plant -specific
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feedwater system perfomance corresponds to the analysis assump--
tions, the feedwater inlet nozzle pressure oscillations will be monitored throughout the design operating flow ranges during the power escalation period following installation of the.preheater modification.. Monitoring will be performed during subsequent periods if necessary to obtain additional or missed data.
Feedwater pump alignment changes will be made during power escala-tion. Transients due to changes in feedwater pump alignment represent the most significant transients on feedwater pressure-oscillation. As such, all significant pressure variations which could affect the fatigue usage factors of the critical. manifold bolts and welds will be monitored. The staff finds the inlet pressure monitoring program acceptable.
The analysis by Westinghouse for steady state pressure fluctuation resulted in the development of curves of allowable peak-to-peak i
pressure oscillations versus frequency.. These were developed for critical modification components most subject to this loading and i
are based on limiting the oscillating pressure stresses at any frequency to the endurance limit for the material.
Acceptance criteria for this test will be established to verify that plant measurements fall within the bounding values used by l
Westinghouse in the analysis of the manifold.
C.-
Radiological Considerations - ALARA Guidelines The staff has evaluated the Utility Design Review Panel's radiological-assessment of the radiation protection measures established by Westinghouse for the Westinghouse Preheat Steam Generator D2/D3 Design Modification, including those measures intended to ensure that doses will be maintained as low as is reasonably achievable ( ALARA). Our assessment is based on the Design Review Panel's (DRP) utilization of the criteria outlined in Regulatory Guide' 8.8,."Information Relevant to Ensuring that Occupational Exposures at Nuclear Power Stations Will Be As Low As Is Reasonably Achievable" in the DRP's radiological assess-ment of the design modification, and their assessments, primarily those provided in Sections 4.4, " Radiological Considerations," Section 5.5,
" Radiological Considerations /ALARA," and Section 6.0, " Summary of the January 1983 DRP Evaluation Report." Infomation provided in other l
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. sections was also considered where it contributed to our assessment of the ALARA features of design, planning, installation, maintenance, and inspection. We have additionally evaluated information specific to the V. C. Summer radiation protection /ALARA program which has pre-viously been submitted in the V. C. Summer Final Safety Analysis Report (FSAR). This has been evaluated and found acceptable by the staff in the V. C. Summer Safety Evaluation Report (NUREG-0717).
Westinghouse has provided information regarding their D2/D3 Design Modification Program in submittals dated December 1982. Additional information has been provided by Westinghouse to the DRP during visits to Westinghcuse facilities and to the DRP and staff during DRP meetings.
The DRP reviewed those actions described by Westinghouse to assure that adequate radiological protection will be provided for workers involved 4
in the modification task, including those measures to be implemented to assure that doses to the workers and the public are ALARA. and has found them to be consistent with Regulatory Guide 8.8.
On this basis, the DRP concluded that adequate consideration has been given to ALARA matters, and that the modification is acceptable from a radiological standpoint. The staff concurs in this conclusion based on the follow-ing:
(1) The personnel on the DRP perfoming the radiological assess-ment were well-qualified professionals in radiation protec-tion; (2) The criteria used by the DRP to conduct their review were i
the same as those used by the staff, namely Regulatory Guide 8.8; l
(3) The Westinghouse Design Modification Package was reviewed l
for completeness by the NRC staff and was found to be comprehensive and detailed enough for a thorough radio-logical asses went; and (4) The DRP Report adequately addresses the key considerations outlined in Regulatory Guide 8.8, providing succinct l
examples of those actions and measures Westinghouse has planned to provide for radiological protection of workers l
and achieve ALARA doses during performance of the modifi-cation.
Westinghouse's proposed D2/D3 Design Modification Program provides for radiation protection /ALARA measures throughout the design and preparation stage, the performance of the modification, and during post-modification j
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,. recovery and operations. The V. C. Summer radiation protection /ALARA program has those features essential for compatibility with the Westinghouse design nodification program, and contains radiation protection /ALARA elements designed to ensure adequate radiological protection for workers and promote ALARA doses on tasks associated with the modification. These proposed measures are consistent with 10 CFR 20.1(c) and Regulatory Guide 8.8, and are therefore, acceptable to the NRC staff for the planned modifications.
The licensee will perfom a summary radiological assessment of the task, as is recommended in C.3.c of Regulatory Guide 8.8, to enable the staff to evaluate the radiological results of the modification and determine if additional or different radiological controls need to be considered. This should include the following:
(1) The collective occupational dose estimate shall be updated weekly.
If the updated estimate exceeds the person-ren estimate by more than 10%, the licensee shall provide a revised estimate, including the reasons for such changos, to the HRC within 15 days of determination.
(2) A final report shall be provided to the NRC within 60 days after completion of the repair. This report will include:
(a) a summary of the occupational dose received by major task, and (b) a comparison of estimated doses with the doses actually received.
By letter dated May 3,1983, the licensee committed to provide a final report to the NRC concerning the radiological dose received during the modification effort including a summary of the occupational dose received by major task and a comparison of estimated doses with the doses actually received. The staff finds this acceptable.
III. EVALUATION The staff has reviewed the V. C. Summer modification to the preheater sections of the steam generators including the feedwater system changes, monitoring program, and radiological considerations. Based on this review, the staff finds that appropriate surveillance measures and remedial action plans have been implemented by the licensee with respect to the Model D3 steam generator tube failure problem.
Therefore, the staff finds that the modification of the V. C. Sunmer steam gener-ators is acceptable and that the modified stean generator can be operated at 100%
of their design capacity.
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IV.
ENVIRONMENTAL CONSIDERATION We have d6temined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any signif-icant environmental impact. Having made this detemination, we have further con-cluded that the amendment involves an action which is insignificant from the stand-point of environmental impact and, pursuant to 10 CFR 51.5(d)(4), that an environ-mental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.
V.
CONCLUSION We have concluded, based on the considerations discussed above, that: (1) because the amendment does not involve a significant increase in the probability or con-sequences of accidents previously considered, does not create the possibility of an accident of a type different from any evaluated previously, and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the comnon defense and security or to the health and safety of the public.
Principal Contributors:
J. B. Hopkins, Licensing Branch Ho. 4, DL J. Rajan, Mechanical Engineering Branch, DE L. Frank, Materials Engineering Branch, DE R. Serbu, Radiological Assessment Branch, DSI l
J. Wermiel, Auxiliary Systems Branch, DSI P. Hearn, Containment Systems Branch, DSI Dated: May 5, 1983 F
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III. EVALUATION-
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The staff has reviewed the V. C. Summer modification to the preh ater sections of the steam generators including the feedwater system changes, m toring program, and radiological considerations.
Based on this review, the s aff finds that appropriate surveillance measures and remedial action plans ave oeen implemented by the licensee with respect to the Model D3 steam generat tube failure problem.
Therefore, the staff finds that the modification of the V C. Summer steam gener-ators is acceptable and that the modified steam generat can be operated at 100%
of their design capacity.
IV.
ENVIRONMENTALCONSIDERATION We have determined that the amendment does not aut/lorize a change in effluent types or total amounts nor an increase in power level 9nd will not result in any signif-icant environmental impact. Having made this d9 termination, we have further con-cluded that the amendment involves an action w 1ch is insignificant from the stand-point of environmental impact and, pursuant t 10 CFR 51.5(d)(4), that an environ-mental impact statement or negative declaratS n and environmental impact appraisal
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need not be prepared in connection with the
,ssuance of this amendment.
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CONCLUSION We have concluced, based on the consider tions discussed above, that- (1) because
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the ame.ndment does not involve a signif ant increase in the probability or con-sequences of accidents previously consi ered, does not create the possibility of an accident of a type different from any valuated previously, and does not involve a significant decrease in a safety marg'n, the amendment does not involve a significant hazards consideration, (2) there is r asonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in com iance with the Commission's regulations and the issuance of this amendment will no be inimical to the common defense and security or to the health and safety of the public.
Principal Contributors:
J. B. H pkins, Licensing Branch No. 4, DL J. Raj n, Mechanical Engineering Branch, DE L. Fr nk, Materials Engineering Branch, DE R. S bu, Radiological Assessment Branch, DSI J. V rmiel, Auxiliary Systems Branch, DSI P.
earn, Containment Systems Branch, DSI l
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